DEMO Concept Development and Assessment of Relevant Technologies. Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor

Transcription

1 FIP/3-4Rb FIP/3-4Ra DEMO Concept Development and Assessment of Relevant Technologies Y. Sakamoto, K. Tobita, Y. Someya, H. Utoh, N. Asakura, K. Hoshino, M. Nakamura, S. Tokunaga and the DEMO Design Team Japan Atomic Energy Agency Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor N. Asakura, K. Hoshino, H. Utoh, 1) K. Shinya, Y. Someya, K. Shimizu, S. Tokunaga, K. Tobita, 2) N. Ohno, 3) M. Kobayashi, 3) H. Tanaka Japan Atomic Energy Agency, 1 Toshiba Nuclear Engineering Services Co. 2 Nagoya University, 3 National Institute for Fusion Science 25 th Fusion Energy Conference (FEC) 2014, St Petersburg October (Friday) am 17(Friday) pm

2 Introduction: Demo concept design and Advanced divertor study DEMO is a bridge from ITER to a commercial reactor, and will demonstrate Electric power generation, Tritium self-sufficiency, Steady-state operation. Breeding blanket and large power exhaust are principal design issues. Design parameters for DEMO have been studied with considering, Medium size (R p > 8m) for full inductive I p ramp up by CS coil (DY CS R CS2 ) Fusion power (P fus < 2 GW) compatible with power handling in divertor. To minimize the development subjects, it is designed by utilizing existing technologies from Tokamaks (ITER, JT-60SA,.) and Nuclear reactor technologies (PWR). Advanced divertor study will provide new options of magnetic configuration. Advanced divertor is produced by driving reverse current in one of the divertor coils Coil currents and number are increased. Super-X divertor image for SlimCS Physics and Engineering issues were investigated in a super-x divertor with a short divertor leg, comparable to the conventional divertor size. Divertor performance was compared in SlimCS (R = 5.5m, P fus = 3GW, I p =16.7MA) in order to compare the previous results in the conv. divertor.

3 F82H Cu alloy Basic concept of divertor FIP/P8-11 K. Hoshino et al. Water-cooling and W mono-block target design: Lower peak heat load (< 5MW/m 2 ) is required for W-target&F82H(RAFM)-cooling tube design. Assessment of reduction in heat load by SONIC Peak heat load at outer target, incl. plasma, surf. recomb., radiation and neutral loading Partial detached Simulation study for R p =5.5 m SlimCS indicated: q target < 10 MW/m 2 is obtained with large radiation fraction (f rad =P rad /P out > 80%, P out = MW) for P fus =1.5-2 GW, suggesting q target < 10 MW/m 2 in larger R p with P fus =1.5 GW and f rad = 70% (q target P out /R p ). Assessment of heat removal capability P fus = 1.5 GW operation reduces dpa/year < 1.5 near the strike-point: W-target&Cu-alloy-cooling tube will be applied at inner and outer targets. Replacement of the divertor target is expected in 1-2 years. Full detached R p =8.2m

4 Assessment of blanket thickness Water cooled solid breeder based on ITER-TBM PWR condition (~300, 15.5 MPa) Be 12 Ti and Li 2 TiO 3 pebbles Overall TBR >1.05 is evaluated for blanket thickness of 0.6m and P fus < 2.0 GW. Blanket a p Plasma r W D BLK Assessment of vertical stability No use of in-vessel coil in DEMO Stabilize plasma by conducting shell, typically at r W /a p ~ 1.35 Vertical stability analysis indicates design elongation of k 95 ~ D SOL D BLK D gap D BLK corresponds to 0.6m for R p =8.2m Shell A=3.2

5 Normalized Performance Absolute Performance Size & Configuration Based on the assessment, possible design/plasma parameter sets are evaluated by systems code (TPC). Key concepts R p > 8 m for full inductive I p ramp up. Operation flexibility from pulse to steady-state P fus ~ 1.5 GW and P gross ~ 0.5 GW based on the assessments of divertor heat removal and overall TBR > 1.05 k 95 = 1.65 for vertical stability with conducting shell. P o s t e r B T max > 12 T based on Nb 3 Sn or Nb 3 Al, S m = 800 MPa Segmented maintenance scheme: Segment RM image for blanket and divertor Analysis of Accident Scenarios: SEE/P5-10 M. Nakamura, et al. Parameters DEMO (Steady state) Ref. ITER R p (m) a p (m) A k q I p (MA) B T (T) B max T (T) P fus (MW) P gross (MWe) Q P ADD (MW) n e (10 19 m -3 ) NWL (MW/m 2 ) HH 98y b N f BS n e /n GW f He

6 1. Concept design study of advanced divertors for DEMO N. Asakura, et al., Trans. Fus. Sci. Tech. 63 (2013) 70. Study showed large current (>100 MAT) is required for the divertor coils outside TFC Divertor coils should be installed inside TFC: interlink-winding Snowflake: Flux expansion (f exp ) increases near SF-null and Connection length (L // ) is times, while f exp div and target wet area are smaller than conv. divertor, appropriate for compact divertor concept. For r mid =1mm Short Super-X: f exp and L // increase along divertor leg radiation and detachment control in divertor. Interlink coil current and number are less than SFD. Snowflake divertor (SFD) Short super-x divertor (SXD) in outer leg I p =16.7MA B t =6.0T

7 2. Conceptual design of Short super-x divertor Interlink divertor coil and the short SXD are arranged under Engineering restrictions: Arrangement of PFCs with 2 interlink-coils: Interlink coils outside neutron shield and vessel. Divertor cassette and its replacement: same height as SlimCS divertor Superconductor interlink design: Nb 3 Al maximum current 25MAT and 1.6m size Magnetic configuration of short-sxd and conventional divertor geometries 2 interlink arrange for SlimCS I p =16.7MA, B t =6.0T Blanket maintenance port n-shield&vessel(0.8m) Divertor maintenance& Exhaust port f SXD = (Y SX-null Y ax )/(Y Div-null Y ax ) = 0.99: max f exp and L // increases to 19 and 2 times than conv. div. Interlink coils

8 Engineering design and issues for Interlink divertor coil Nb 3 Al Superconductor is preferable for Interlink divertor coil than Nb 3 Sn: (1) React and Wind (2) Stress analysis (< 250MPa): lower load ratio (<50%) of allowable stress (500MPa) Design issues and Development: SC filament is reduced from 60 to 1mm (equivalent to Nb 3 Sn) to decrease AC losses. EM-force on IL-coil (-23 MAT) becomes MN under average Br (0.67T) additional load on TFCs support of IL-coil is necessary. Winding image of Nb 3 Al conductor: Superconductor coil is inserted though TF coils, and is winded by rotating SC bobbin. Interlink superconductor is designed, based on ITER poloidal coil conductor: Supercritical helium flow (f5mm) 2nd IL coil(#8) Conduit Insulator Seperconductor (4.6kA) 1st IL coil (#9) 25 MAT corresponds to coil size of 1.6mx1.6m H. Utoh, et al. Fusion Eng. Des. 89 (2014) 2456.

9 3. Divertor plasma simulation of short SXD by SONIC code SONIC simulation for short SXD : Input parameters are the same as conventional divertor: P out = 500 MW, n i =7x10 19 m -3 at core-edge boundary, c i = c e = 1 m 2 s -1, D = 0.3 m 2 s -1 :same as ITER simulation Radiation power loss is increased by Ar seeding at the same total radiation fraction (P rad /P out ) = 0.92 as SlimCS divertor analysis (IAEA FEC2012). Radiation power is increased in the divertor, compared to reference divertor Impurity retention is improved. SOL calculation width (r mid =2.2 cm) core-edge bounary Note: n Z /n i (SOL) ~ 1.5% in short-sxd and LL, while ~2% in reference. N. Asakura, et al. Nucl. Fusion, 53 (2013) Full detachment (T e =T i ~1eV) is produced at both targets.

10 Detachment is produced near SX-null in short SXD Radiation is enhanced near the SX-null (along the separatrix). At the same time, high temperature plasma (>100 ev) is maintained near the SX-null (in Poster). Radiative area in the poloidal direction is narrow due to longer fieldline length. P rad Short SX divertor Conv. long-leg Conv. reference SX-null full detach Partial detach Maximum total heat load ~10 MWm -2 in the full detached divertor (T e = T i ~ 1eV) Surface recombination is dominant near the separatrix, due to large ion flux. Short SX divertor Conv. long-leg Conv. Ref.

11 Summary: Demo concept design and Advanced divertor study DEMO concept is considered through the assessment of relevant technologies. R p > 8.0 m for full inductive I p ramp up by CS coil Operation flexibility from pulse to steady-state P fus = 1.5 GW and P gross ~ 0.5 GW are foreseeable from the viewpoints of Divertor heat removal capability and tritium self-sufficiency in blanket. By considering above and other assessments, DEMO concept design study shows, f rad = 70% will be compatible with P fus =1.5 GW, R p > 8.0 m and partial use of Cualloy as cooling pipe near high q target and lower dpa/year region Water cooled solid breeder blanket with its thickness of 0.6 m for TBR > 1.05 k 95 = 1.65 for vertical stability with conducting shell Segmented maintenance scheme. Re-use & recycle of components. B max T > 12 T is achieved, based on both Nb 3 Sn or Nb 3 Al, S m = 800 MPa Advanced divertor study will provide new options of the divertor configuration. Physics and Engineering issues of Short-SXD has been studied in SlimCS: Interlink divertor coils are required: Nb 3 Al SC is preferable for React&Wind SC filament size should be reduced, and IL-coil support for EM-force is required. f exp and L // to target are increased along the divertor leg: max. 19 and 2 times. Power handling has been investigated by SONIC for P FP = 3GW reactor(p out =500MW) Radiative area is narrow poloidally, and efficient to produce full detachment. Surface recombination is dominant near the separatrix due to large ion flux. Conv. divertor is the first choice: Advantages and issues in adv. divertors are studied as alternative. P o s t e r